NUCLEAR AND INDUSTRIAL ENGINEERING (ITALY)

The EBR-II benchmark specifications and designs were used to develop the thermal hydraulic model of the reactor. The RELAP5-3D system thermal hydraulic code was used for preparing the nodalization.

7.8.1. Geometry/discretization

A detailed nodalization reproducing each geometrical zone of the reactor was developed.

Basically, the model can be divided into two parts: the core, which consists of the central core, inner blanket and outer blanket regions, and the coolant system, which includes the pool and the remaining part of the primary sodium circuit (i.e. high and low pressure piping, pumps, inlet plena, upper plenum, Z-Pipe, IHX and the secondary side).

The whole core region consists of 96 channels, representing all 10 types of subassemblies used in the reactor, and two bypasses. The core was divided into 16 rows, according to the real geometry of the EBR-II core. The first 6 rows that represent the central core region were modelled individually (1 subassembly per channel) with 81 channels, except for the safety/control rods, which were combined into one channel. Rows 7 to 16 consist of reflector or blanket subassemblies, and they were modelled with one channel per type of subassembly in each row. Each channel is made up of 36 thermal hydraulic volumes, where the active part of the reactor core has 24 volumes. From both the hydraulic and the thermal point of view, the core is divided into two zones: the central core region simulating the driver subassemblies and the external core region representing the reflector and blanket subassemblies.The heat structures for each subassembly in the central core region consist of:

(a) 1 heat structure component used to simulate the active part of the fuel pins;

(b) 1 heat structure component used to model the non-active part of the fuel pins and the steel rods (if present);

(c) 1 heat structure component used to model the gas plenum;

(d) 6 heat structure components to represent each edge of the subassembly walls.

The pin power profile imposed on the active heat structures was assumed to be flat and constant along the entire active length. In the second phase of the benchmark, an axial power profile was also assumed below and above the active part of the fuel, to take into account gamma heating. The heat structures in the inner and outer blanket regions were modelled with two heat structure components for each row, one to simulate the internal rods and the other one to represent the subassembly walls.

The pool was initially modelled with three parallel pipes connected above and below with two branch components and with all the nodes connected radially to simulate the mixing of sodium among them. In the second phase of the benchmark the three pipes were replaced with a cylindrical multidimensional component having 2 radial meshes, the internal one coinciding with the reactor vessel cover; 3 azimuthal meshes, thermally linked to the pumps or the IHX;

and 72 axial meshes, to preserve the sliced approach adopted in the nodalization. The region of the 3-D pool occupied by the reactor vessel cover was blocked.

The pumps were modelled with a PUMP component, with the homologous curves implemented based on the specifications provided for the benchmark. The high and low pressure piping was modelled by one dimensional components (i.e. PIPE and BRANCH), as were the inlet plena, which were modelled using two sets of BRANCH components: one set for the high pressure plenum that feeds the central and expanded core regions (the first 7 rows), and the other set for the low pressure inlet plenum that feeds the outer blanket region.

Also the upper plenum and the Z-Pipe were modelled by one dimensional components: 6 BRANCH components simulated the upper plenum in a fictitious 3-D model and a PIPE component modelled the Z-Pipe, connected by two single junctions to the upper plenum and to the IHX.

The intermediate side of the EBR-II reactor was represented with the IHX, modelled as a counter-current flow-type heat exchanger. Both the primary side and the secondary side of the IHX were modelled by one dimensional components (i.e. PIPE and BRANCH). The boundary conditions of the secondary system were set according to the reactor design using TMDPVOL and TMDPJUN components.

7.8.2. Nuclear and thermo-physical data/correlations

From the hydraulic part of the model the standard RELAP5-3D thermodynamic properties of sodium were used. The thermal properties of the heat structures were derived from the benchmark specifications. No specific correlation was developed for the simulation.

7.8.3. Thermal hydraulics methods and models 7.8.3.1. Code(s) used

The EBR-II nodalization was developed using the RELAP5-3D V4.1.3 system thermal hydraulic code.

7.8.3.2. Basic method

The following general rules, among others, were adopted during the development of the RELAP5-3D nodalization of EBR-II:

(a) The ratio between the volumes of two adjacent nodes shall be between 0.5 and 2;

(b) The ratio between the lengths of two adjacent nodes shall be between 0.5 and 2;

(c) To use a standard set of code options;

(d) To use more than nine mesh points for simulating the heat structures of the fuel bundles;

(e) To adopt the “slice technique” approach in order to improve the capability of the code and of the nodalization to simulate phases of transients involving natural circulation phenomena.

Regarding the last item, given the fact that the density of liquid sodium is much higher than water, a sliced approach was necessary for avoiding any kind of oscillations in the code calculation and unrealistic pressure differences among parallel flow paths. The sliced approach is a nodalization technique consisting of dividing the hardware in parallel slices in order to have the centres of each node in parallel pipes at the same elevation position. This is a good practice to better reproduce phenomena connected with natural circulation, where small gravitational head differences play a significant role.

7.8.3.3. Model

As default, the following RELAP5-3D models were adopted for the nodalization of EBR-II:

(a) The non-equilibrium (unequal temperature) calculation was used;

(b) The non-homogeneous (two-velocity momentum equations) option was activated;

(c) Use of momentum flux in both the ‘to volume’ and the ‘from volume’;

(d) The vertical stratification model was used for the volume;

(e) The choking model was adopted (if a choked flow condition is predicted by the code);

(f) The wall friction effects were computed along the x-, y- and z- coordinates of the volume.

7.8.4. Blind results 7.8.4.1. SHRT-17

After achieving acceptable steady state conditions, the blind transient calculation was performed. Starting from full power and flow, both the primary loop and intermediate loop coolant pumps were simultaneously tripped and the reactor was scrammed to simulate a protected loss of flow accident. In addition, the primary system auxiliary coolant pump, that normally had an emergency battery power supply, was turned off. The reactor core power and the intermediate side boundary conditions were imposed according to the benchmark specification.

At this stage of the benchmark, the EBR-II primary pumps were modelled identically (as can be seen from FIG. 134). During the MCPs coastdown (up to about 10 s) the cladding and the outlet coolant temperature decreased in both instrumented subassemblies (see FIG. 135 for XX09 and FIG. 136 for XX10) due to the sharp decrease in the nuclear fission power. During the transition from forced to natural circulation (between about 10 and 100 s), the imbalance between the total core power and the energy removed from the primary coolant caused a rapid increase in the cladding temperatures and a somewhat less rapid increase in the coolant

temperatures. Note that in instrumented subassembly XX10 the cladding temperatures increased in two distinct steps (see FIG. 136). As can be seen in FIG. 137, the calculated mass flow rate in subassembly XX10 became negative for about 130 seconds. Since the power generated in this subassembly is quite low compared to the others, the temperature increase is due mainly to the decrease of the mass flow rate and occurred in parallel with the flow reversal (i.e. when the mass flow rate became negative and vice versa). When natural circulation was fully established (after about 100 s) the total core power was efficiently removed in all subassemblies and the coolant and cladding temperatures decreased.

FIG. 134. SHRT-17 primary pump mass flow rates, blind results.

FIG. 135. SHRT-17 coolant and cladding temperatures of XX09, blind results.

FIG. 136. SHRT-17 coolant and cladding temperatures of XX10, blind results.

FIG. 137. SHRT-17 mass flow rates in XX09 and XX10, blind results.

7.8.4.2. SHRT-45R

N.IN.E. did not perform an analysis of SHRT-45R.

7.8.5. Final results, data comparisons 7.8.5.1. SHRT-17

The final results of the RELAP5-3D simulation of SHRT-17 are reported below. Among the various improvements developed for the final simulation, discussed in more detail in Section 7.8.5.2, the most significant concerns are:

(a) The mass flow rate: the primary pumps were modelled as separate pumps, with the pump speeds updated according to the input specifications, and the energy loss coefficients in the subassemblies were improved, taking into account their dependence on the Reynolds number;

(b) The axial power distribution: a power source was added below and above the active part of the fuel, to take into account gamma heating. This will be discussed in more detail in Section 8.3, where sensitivity analyses are covered.

Following the primary and intermediate pump trips and the reactor scram, the calculated mass flow rates decreased rapidly and the coolant and cladding temperatures started to increase.

During the pump coastdown, the calculated mass flow rate in instrumented subassembly XX09 remained a little bit higher than the experimental data (see FIG. 138). This affected the coolant and cladding temperatures in the whole subassembly. Indeed, both the coolant temperatures below (FIG. 139) and above (FIG. 140) the active part of the rods and the cladding temperatures at the middle and at the top of the core (FIG. 141) were slightly lower than the experimental data. These small differences became negligible in the latter portion of the transient because the mass flow rate reached the correct value. It should be noted that the flowmeter temperatures, where the gamma heating occurred, were predicted well qualitatively by the code simulation.

Conversely, in instrumented subassembly XX10, the mass flow rate during the pump coastdown (FIG. 142) reached a slightly lower value compared to the experimental data, followed by a faster increase before it stabilized at the correct value. The effect of this flow prediction can be seen especially in FIG. 143, where the cladding temperatures at the middle and at the top of the core are shown. In both curves, the temperatures reached slightly higher peak values, followed by a bit faster decrease before reaching the correct value during the last half of the transient.

Regarding the coolant temperatures outside the active part of the core, since the power to flow ratio of subassembly XX10 is lower compared to the remaining subassemblies, the gamma heating and the heat conduction with the adjacent subassemblies play an important role in the temperature trends. Below the core, the predicted coolant temperature (see FIG. 144) remained a little bit lower than the experimental data during the pump coastdown because the gamma heating was not modelled well in the heat structure (in this case, a certain percentage of the subassembly power was used to model gamma heating, based on the analysis of instrumented subassembly XX09, see Section 8.3). At the subassembly outlet and in the guide thimble annulus (see FIG. 145) the coolant temperature decreased early compared to the experimental data after the pump coastdown because the gamma heating, and therefore also the heat conduction from the adjacent subassemblies, was slightly underestimated. However, during the latter half of the transient, the coolant and cladding temperatures were predicted well by the code.

FIG. 138. SHRT-17 XX09 mass flow rate, final results.

FIG. 139. SHRT-17 XX09 flowmeter temperatures, final results.

FIG. 140. SHRT-17 XX09 coolant temperatures, final results.

FIG. 141. SHRT-17 XX09 cladding temperatures, final results.

FIG. 142. SHRT-17 XX10 mass flow rate, final results.

FIG. 143. SHRT-17 XX10 cladding temperatures, final results.

FIG. 144. SHRT-17 XX10 flowmeter temperatures, final results.

FIG. 145. SHRT-17 XX10 coolant temperatures, final results.

7.8.5.2. Model improvements

As mentioned before, several improvements were implemented in the RELAP5-3D model during the benchmark. Initially, the reactor core was modelled with 25 channels total, collapsing 16 rows into 9 rows. The first 7 rows were modelled using one PIPE component for each type of subassembly located in each row. The remaining part of the core was split into two regions, one region formed by merging rows 8 through 11, and the other region

formed by merging rows 12 through 16, always keeping the different types of subassemblies separated. Also, the heat structures inside the core were simulated following the same approach adopted for the hydraulic part, using three different heat structures for each channel:

one modelling the active part of the core, one representing the gas plenum, if any, and the last one simulating the hexagonal tube and other non-active heat structures (i.e. upper and lower shield, stainless steel rods, guide thimble) if present.

In the last version of the RELAP5-3D model, the whole core consisted of 96 channels that represent all 10 types of subassemblies used in the reactor, plus two bypasses. The first 6 rows were modelled separately (1 channel per subassembly) with 81 channels, except for the safety/control rods, which were combined into one channel. Rows 7 through 16 were modelled with one channel per type of subassembly in each row. In addition, for the two instrumented subassemblies (XX09 and XX10), each guide thimble flow region was modelled with a PIPE component. Also the number of heat structures was increased within the hydraulic channels, especially in the core central region. In particular, the subassembly walls were modelled with 6 heat structure components to represent each edge of the hexagonal tube and thermally connect with the adjacent subassemblies.

Regarding the power distribution, the initial power was updated to the value recorded for SHRT-17 (equal to 57.29 MW instead of 59.97 MW of run 129 C). The pin power profile was kept flat and constant along the active length, but an axial power profile was implemented below and above the active part of the fuel, to take into account gamma heating.

Another significant improvement is related to the cold pool. The three parallel pipe components were replaced with a cylindrical multidimensional component having 2 radial mesh cells, the internal one coinciding with the reactor vessel cover; 3 azimuthal mesh cells, thermally linked to the pumps or the IHX; and 72 axial mesh cells, to preserve the sliced approach adopted in the nodalization. The region of the 3-D pool occupied by the reactor vessel cover was blocked.

The velocity of pump #2, initially set equal to that of pump #1, was modified to match the benchmark specifications. Also the energy loss coefficients were modified to improve the prediction of the mass flow rate in the subassemblies, taking into account their dependence on the Reynolds number.

The heat structure of the Z-Pipe was modified by adding a double layer of stainless steel with stagnant sodium that filled the annular region. Also, the IHX was changed by refining the heat structure simulating the tube bundle and modifying the intermediate side in order to have both the inlet and the outlet of the intermediate side pipes at the top of the IHX.

Finally, the leakage flow paths were also modified to be consistent with the benchmark specifications. Initially, they were all collapsed and located in the inlet plena. In the final model, the leakage paths were placed in the correct locations throughout the primary sodium circuit, including the three clearance flow paths through the core.

7.8.6. Neutronics methods and models

N.IN.E. did not participate in the neutronics benchmark.